ML19257A081

Amend 24 to License DPR-69,revising App a of Tech Specs to Increase Measurement/Calculational Uncertainties for Peaking Factors

ML19257A081
Person / Time
Site:	Calvert Cliffs
Issue date:	12/11/1979
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19257A079	List: ML19257A078 ML19257A081 ML19257A083 ML19257A086
References
NUDOCS 8001020070
Download: ML19257A081 (13)

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BALTIM0RE GAS & ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR 70WER PLANT, UNIT NO. 2 AMENDMENT TO FACILITY'0PERATING LICENSE Amendnent No..24 License No. DpR-69 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for anendment by Baltimore Gas and Electric Company (the licensee) dated August 27, 1979 as supplemented October 1,1979, complies with the standards and requirments of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment.will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment !s in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

1665 150 8001020 O

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69 is hereby amended to read as follows:

2, Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 24, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FORTHENUCLEARREGULATORYCbMMISSION G2hd W Gl]'

~

Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 11, 1979 1665 151 e

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ATTACHMENT TO LICENSE AMENDMENT NO. 2 4 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET N0. 50-318 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresporiding overleaf pages are also provided to maintain document completeness.

Pages I

1-6 3/4 2-2 3/4 2-4 B 3/4 2-1 1665 152

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS D e f i n e d T e rm s..............................................

1-1 Th e r na l Po we r..............................................

1-1 R a t e d T h e rma l P owe r........................................

1-1 Operational Mode...........................................

'l-1 Action.....................................................

1-1 h e ra b l e - O p e ra b i l i ty....................................

1-1 Reportable Occurrence......................................

1-2 Co n t a i nmen t I n te g ri ty......................................

1-2 Channel Calibration........................................

1-2 Channel Check..............................................

1-3 Ch a n nel Fu ncti on al Te s t....................................

1-3 Core Alteration............................................

1-3 Shutdown Margin............................................

1-3, Identified Leakage........................................

1-4 Un i den ti fi ed Lea k a ge.......................................

1-4 Pressure Boundary Leakage..................................

1-4 Controlled Leakage.........................................

1-4 A z i mu t h a l Powe r T i l t.......................................

1-4 Dose Equivalent I-131......................................

1-4 E-Ave rage Di si n tegra ti on Ene rgy............................

1-5 Staggered Test Basis.......................................

1-5 Frequency Notation.........................................

1-5 Axial Shace Index..........................................

1-5 Uneodded Planar Radial Peaking Factor - F 1-5 xy........

Reactor Trip System Response Time..........................

1-6 Engineered Safety Feature Response Time....................

1-6 Physics Tests.................

1-6 Unro'dded Integrated Radial Peaking Factor - F.............

1-6 r

CALVERT CLIFFS - UNIT 2 I

Amendment No. 9, JS,q 4 1665 153

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core................................................

2-1 Reactor Coolant System Pressure.............................

2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS' R ea c to r T r i p S e tpo i n ts......................................

2-6 BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core................................................

B 2-1 Reac tor Coola nt System Pressure.............................

B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints......................................

B 2-4 g665 154 CALVERT CLIFFS - UNIT 2 II

DEFINITIONS f - AVERAGE DISINTEGRATION ENERGY 1.20 f shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma. energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

A test schedule for n systems, subsystems, trains or other a.

designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train or other designated component at the beginning of each subimterval.

FREQUENCY NOTATION 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

AXIAL SHAPE INDEX 1.23 The AXIAL SHAPE INDEX (Y ) is the power level detected by the lower E

excore nuclear instrument detectors (L) less the power level detected by the upper excore nuclear instrument detectors (U) divided by the sum of these power levels. 'The AXIAL SHAPE INDEX (Y ) used for the trip and r

pretrip signals in the reactor protection system is the above value (Y )

modified by an appropriate multiplier (A) and a constant (B) to determkne the true core axial power distribution for that channel.

Yg = AYE+B YE=

UNRODDED PLANAR RADIAL PEAKING FACTOR - Fxy 1.24 The UNRODDED PLANAR RADIAL PEAKING FnCTOR is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding tilt.

CALVERT CLIFFS-UNIT 2 1-5 Amendment No. 9 1665 155

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

ENGINEERED SAFETY FEATURE RESPONSE TIME l.26 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setooint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures Yeaf.1 their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the funda-mental nuclear characteristics of the reactor core and related instrumen-tation and 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

UNR00DED INTEGRATED RADIAL PEAKING FACTOR - Fr 1.28 The UNRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average pin power in an unrodded core, excluding tilt.

1665 156 CALVERT CLIFFS-UNIT 2 1-6 Amendment No. 9, JB, 2 4

3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3. 2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1.

APPLICABILITY: MODE 1.

ACTION:

With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a.

Restore the linear heat rate to within its limits within one hour, or b.

Be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.

4.2.1.3 Excore Detector Monitoring System - The excore detector moni-toring system may be used for monitoring the core power distribution by:

a.

Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6.

b.

Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.

1665 157 CALVERT CLIFFS-UNIT 2 3/42-1 Amendment No. 9, 18

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) c.

Verifying at least once per 31 days that the AXIAL SHAPE INDEX is maintained within the limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN where:

1.

M is the maximum allowable THERMAL POWER level for the existing Peactor Coolant Pump combination.

2.

N is the maximum allow 9ble fraction of RATED THERMAL POWER as determined by the F curve shown up Figure 3.2-3 of

• Y Specification 3.2.2.

4.2.1.4 Incore Detector Monitorina System - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a.

Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE 1.

b.

Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the settino of these alarms:

1.

Flux peaking augmentation factors as shown in Figure 4.2-1, 2.

A measurement-calculational uncertainty factor of 1.07, l

3.

An engineering uncertainty factor of 1.03, 4.

A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thermal expansion, and 5.

A THERMAL POWER measurement uncertainty factor of 1.02.

1665 158

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CALVERT CLIFFS-UNIT 2 3/4 2-2 Amendment No. 6, 9, 76, JS,9.4

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g FIGURE 3.2-2 Linear Heat Rate Axial Flux Offset Control Limits CALVERT CLIFFS - UNIT 2 3/4 2-4 Amendment NO. 9,JS24 1665 160

3/4.2 POWER DISTPIBUTION LIMITS BASES 3/4.2.1 LINEAR MEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.

Either of the two core power distribution monitoring systems, the Ercore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.

The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.

In conjunction with the use of the excore monitoring system and in establishing the AXI AL SHAPE INDEX limits, the following assumptions are made:

1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.3 are satisfied, and 4) the TOTAL RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.

The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alams which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1.

The setpoints for these alarms include allowances, set in the conservative directions, for 1) flux peaking augmentation factors as shown in Figure 4.2-1, 2) a measurement-calcelational uncertainty factor of 1.070*, 3) an engineering uncertainty factor of 1.03, 4) an allowance l

of 1.01 for axial fuel densification and thermal expansion, and 5) a THERMAL POWER measurement uncertainty factor of 1.02.

3/4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING T

T FACTORS - F AND F AND AZIMUTHAL POWER TILT - T xy r

q T

The limitations on F and T are provided to ensure that the assumptions used in the aMlysis for establishing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operation at yhe various allowable CEA group insertion limits.

are provided to ensure that the assumptions The limitations on F used. in the analysis # and Testablishing the DNB Margin LCO, and Themal CALVERT CLIFFS - UNIT 2 B 3/4 2-1 Amendment No. 9 JS 3 4 1665 161

POWER DISTRIBUTION LIMITS BASES Margin / Low Pressure LSSS setpoints remain valid durjng o eration at the various allowable CEA group insertion limits.

If F F or T exceed their basic limitations, operation may continue undd/,th addi9ional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The value of T that must be used in the equation F

=F (1 + Tq) q and F *fr (1 + T ) is the measured tilt.

r q

T T

The surveillance requirements for verifying that F p and T within their limits provide assurance that the actual ylYu,es 9f FT4hye 7

T r

and T do not exceed the assumed values. Verifying F and F afNr each 9uel loading prior to exceeding 75% of RATED THEMAL POWER provides additional assurance that the core was properly loaded.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum CE-1 calculated DNBR of 1.19 throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of tt. flow indication channels with neasured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

1665 162 CALVERT CLIFFS - UNIT 2 B 3/4 2-2 Amendment No. 7, M. 18 3,

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